The present invention relates to a dissolver for removing nuclear fuel materials from fuel element segments during reprocessing of irradiated nuclear fuels.
In such apparatus, for reasons of criticality, the structural parts are made of a material which is a neutron absorber such as, for example, hafnium, and the dissolver is composed of a dissolving vessel into which a dissolving basket containing the fuel element sections which are to be subjected to the dissolving process can be placed so that fluid can flow therethrough. The fuel elements of nuclear reactors in the majority of cases are composed of the actual nuclear fuel and a metallic protective sleeve or shell.
At the beginning of such a reprocessing procedure, the irradiated reactor fuel elements are initially cut up mechanically into short fuel rod segments. These fuel rod segments, together with the structural components of the fuel elements, which include spacers and head and foot pieces, drop into the dissolving basket which is disposed in the dissolving vessel. In the dissolver, the nuclear fuel material is dissolved out of the fuel rod segments by means of boiling nitric acid. Upon completion of the dissolving process, the nitric acid-containing fuel material solution is extracted, the empty sleeves are washed with fresh acid and then the basket together with the empty sleeves and the other structural material is removed from the dissolving vessel, is washed with fresh water and then the basket is emptied.
The fuel material solution is subjected to further chemical processing for separating the reusable nuclear fuel material, preferably uranium and plutonium as well as possibly thorium, from the fission products. The leached sleeves and the structural material constitute solid radioactive waste and are treated and stored accordingly.
During the dissolving of the fuel materials, the presence of fissionable material, preferably uranium and/or plutonium, poses a particular problem. This involves the danger of establishment of a "critical state" in the system in which a self-sustaining nuclear chain reaction could take place. For this reason measures must be taken which prevent such critical excursion under any circumstances.
This problem has heretofore been solved by constructing the dissolver to have "geometrically critically safe" external dimensions, i.e. its dimensions were limited to values which, under consideration of the amount of fissionable material contained in the fuels to be processed, lie below the minimum citical dimensions. However, this greatly limits the volume, and thus the capacity, or fuel throughput, of the dissolver. Such dissolvers can thus be used economically only for small reprocessing systems or for nuclear fuel materials having a low content of fissionable material, such as, for example, those from heavy water natural uranium reactors.
In order to have available larger capacity dissolvers, for example for system throughputs of several tons of uranium per day, for nuclear fuel materials from modern power reactors containing greater amounts of fissionable material, e.g. light water reactors with enriched uranium oxide or with plutonium oxide/uranium oxide as the fuel material, it is known in the art to add a "soluble neutron poison", i.e. a dissolved substance having a high capture cross section for neutrons, particularly gadolinium nitrate, to the dissolving acid. However, this process has the drawback that the dissolved neutron poison is lost in the subsequent chemical reprocessing procedure together with the highly active waste and cannot be recovered. Due to the high cost and limited availability of soluble neutron poisons as well as the highly active waste connected with its use and the problems of further processing such waste, this process is very uneconomical.
Economical operation is made even more difficult by the fact that the constant presence of a soluble neutron poison in the dissolving acid must be continuously verified, involving high monitoring expenses and entailing extensive use of automated control instruments so as to exclude the possibility of a criticality accident.
It has also already been proposed to use hafnium in the reprocessing systems for the purpose of heterogenous poisoning.
The prior art dissolvers for large systems, in which head and foot portions and spacers of the fuel rod bundles are also introduced into the basket, are designed to have a diameter of 70 cm and a height of 5 m. They pose great handling problems to their users and must additionally be loaded with the above-mentioned soluble and nonextractable neutron poisons, which must be continuously added and which adversely influence the highly active waste as salt formers, i.e. they limit the possible concentration by salts which are inherently inactive.
In summary, it can be stated that large reprocessing systems, those with capacities of the order of magnitude of 1500 tons per year, require either too many geometrically safe dissolvers, that is many mechanical devices subject to malfunctions and requiring many operations, or the use of soluble neutron poisons which present the above-mentioned drawbacks.